ML20082P369

Forwards Evaluation of Emergency Feedwater Sys Capacity Vs Generic Westinghouse Analysis in Support of ATWS Rule,Per NRC Questions During 910710 Telcon

ML20082P369
Person / Time
Site:	Seabrook
Issue date:	09/04/1991
From:	Feigenbaum T PUBLIC SERVICE CO. OF NEW HAMPSHIRE
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NYN-91136, NUDOCS 9109100304
Download: ML20082P369 (7)

Text

- _ _ _ _ - _ _ _

e New Hampshire Y

khh Ted C. Feigenbovm Presdent and CMel becutive Of ficer NYN 91136 September 4,1991 i

United States Nuclear Regulatory Commission i

Washington, D.C.

20555 Atter' ion:

Document Control Desk

References:

(a)

Facility Operating License No. NPF 86, Docket No. 50 443 (b) 10CFR50.62, Requirements for iteduction of Itisk From Anticipated Transients Without Scram (ATWS) Events for Light Water + Cooled Nuclear Power Plants S ubject:

Emergency Feedwater Pump Capacity Versus Generic Analysis in Support of l

the ATWF Rule Gentlemen:

In resronse to questions from the NRC Staff during a July 10, 1991 telephone call, New Hampshire Yankee (NilY) is providing in the Enclosure an evaluation of the Seabrook Station Emergency Feedwater System capacity versus the generic Westinghouse analysis provided in support of the ATWS rule.

New !!ampshire utilites an ATWS Mitigation System Actuation Circuitry (AMSAC) to meet the requirements of 10CFR50.62. The Seabrook Station Emergency Feedwater Syst r..

is comprised of one nmtor driven and one turbine driven pump, both of which are actuated by an AMSAC signal.

The evaluadon provided in the Enclosure demonstrates that the Seabrook Station design is in corupliance with 10CFR50.62 and the Rute basis as provided in SECY-83 293, Sheuld you have any qu:si ens regarding this matter pleast contact Mr. James M.

Petchel, Regalatory Complianu Manger, at (603) 474-9521, extension 3772.

Very truly yours, hy

"~

Ted C. F' genbaum T CF:.IM P/act 9109100304 910904 FDR ADOCK 05000443 P

PDR r

000138 New Hampshire Yonkee Division of Public Service Company of New Hampshire k

P.O. Box 300

• Seabrook, NH 03874

• Telephone (603) 474 9521

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- Uni'. ;o

s Nuclear Regulatory Commission September 4,1991 Attenti-Document Control Desk Page two ec

Mr. -Thomas T. Martin Regional Administrator -

United States Nuclear Regulatory Commission' Region i 475 Allendale Road King of Prussia, PA 19406-

Mr.'Gordon E. Edison, Sr. Project Manager Project Directorate i 3 i

Division of Reactor Projects U.S. Nuclear Regulatory Commission

-Washington, DC 20555 Mr. Noel Dudley NitC Senior Resident inspector P.O. Box 1149

' Seabrook, NH 03874 s

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I New llampshire Yankee September 4,1991 ENCL.OSilRTI 1 TO NYNg1_2h l

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SEABROOK STATION EMERGENCY FEEDWATER CAPACITY VERSUS GENERIC ANALYSIS IN SUPPORT OF Tile ATWS RULE 1.0.

INTR ODUCTION in support of the design of the Anticipated Transient Without Scram (ATWS) Mitigation Systems Actuation-Circuitry (AMSAC) for Seabrook Station, a calculation was performed in J uly, 1987, Th'e calculation included a comparison 'of pertinent Seabrook Station design

. parameters versus the generic reference plants analyzed in WCAP 8330 (Reference 1) and in a -letter from' T, M, Anderson on Westinghouse to the NRC (Reference 2).

The comparison included the use of sensitivity studies and equipment failure cases from Reference 1

2 in the evaluation of the design differences between Seabrook Station and the reference i

4 loop plant with Model F steam generators (SGs), The calculation demonstrated that with the installation of AMSAC, Seabrook Station would meet the intent of the ATWS Rule, r

10CFR50.62 and the rule basis provided in SECY 83 293-(Reference 3),

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-2.0' ATWS OVERPRESSURE LIMIT

}.

The Scabrook Station calculation, as does the NRC analysis in Reference 3, assumes that the ATWS overpressure occurs if the pressure limit corresponding 'to the ASME Boiler and Pressure Vessel Code Level C service limit (3200 psig for Westinghouse plants) is exceeded, i

Exceeding 3200 psig is conservatively equated with core damage, 3.0 LIMITING ATWS EVENTS FOR SEABROOK STATION

)

i-In Westinghouse letter NS-TMA 2182 (Reference 2) the two limiting pressure events analyzed were the Loss of Load and/or Turbine Trip (LOL) and the Complete Loss of Normal Feedwater Flow (LOFW). The Seabrook Station calculation included a plant comparison for I-both of these events,

.4.0 SEABROOK STATION VERSUS GENERIC REFERENCE PLANT The peak pressures in the limiting ATWS events for Scabrook Station are discussed below, p

4,1 Four-Loon Plant With Model F Steam Generators l

The 4 loop reference plant with Model F SGs provides the starting point for the it comparison. ' The Seabrook Station core geometry is identical to the reference plant, l

so the fuel heat transfer studies are-not applied to Scabrook Station.

Steam Gecerator design pressure, Reactor Coolant System volume, initial pressurizer water i

level, main feedwater enthalpy, and SG initial inventory are also identical, so these I

. studies also are not applied to Seabrook Station. From Reference j

Peak pressure = 2902 psia for the Loss of Load (LOL) j j.

Peak pressure = 2830 psia for the Loss of Feedwater (LOFW) i l

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4.2 Reactor Power I.evel The Seabrook Station rated core power is 3,411 MWt compared to 3,427 MWt for the reference planti Frota the Reference 2 sensitivity studies, this approximate 0.5% -

reduction in-core power yields the following decrease in peak pressures:

i.

Delte P ='.10 psi for the LOL Delta P = 3 psi for the LOFW 4.3 Pressurirer Sorav

~

4 The reference plant analysis did not take credit for pressurizer spray. The affect of mitigating equipment on - ATWS risk-is discussed separately in Section 5.0 below.

From the Reference 2 ' sensitivity studies, credit for pressurizer spray yields-the following decreases in peak pressures:

L I

Delta P

• 11 -psi for the LOL Delta P = -6 psi for the LOFW 4.4 AMSAC The reference plant analysis (Reference 2) assumed that auxiliary feedwater (AFW) flow was initiated at 60 seconds into the transients. In the LOL cvent, a turbine trip is part of the initiating event. In the LOFW event, the turbine was assumed to be tripped at 30 seconds into the transient. The Reference 2 sensitivity studies showed

-that increasing the turbine trip delay to 60 seconds and initiating AFW flow at 120 seconds will increase the peak transient pressures by:

Delta P = + 134 psi for the LOL Delta P = + 165 psi for the LOFW For Seabrook Station, the. AMSAC trip setpoint, timer delay, and allowances for uncertainties were selected and-considered to ensure Emergency Feedwater System-

. (EFW) initiation.before 120 seconds and a turbine trip before 60 seconds. The above penalties in peak pressure are therefore conservatively applied to Seabrook Station.

4.5 Emercency Feedwater Flow Rate The reference plant analysis (Reference 2) assumed ~an AFW flow rate of 1760 gpm.

The Seabrook Station EFW system can deliver a minimum of 950 gpm from a combination of one motor driven and one turbine driven pump.

130th pumps are actuated by AMSAC during an ATWS event. The affect of the lower EFW flow rate for Seabrook Station is determined from the AFW flow sensitivity study and the AFW pump failure case in Reference 2.

The penalty is:

Delta P = + 59 psi for the LOL Delta P = + 28 psi for the LOFW WCAP-11993 (Reference 4) shows that even with a 50% reduction in AFW flow rate (failure of a 100% capacity turbine driven AFW pump), the result is a "no core 2

i

damage state".

This is also true for Seabrook Station.

With the conservative assumption that one EFW pump ddivers a minimum of 710 gpm:

Delta P = + 18 psi for the LOL Delta P = + 13 psi for the LOFW 4.6 Cqmpatisen Summary Descriotion Value I.OlJI 0FW Peak pressures for 4 loop plant with 2902/2830 Model F SGs, psia Effect of core power (3411 MWt versus

-10/ 3 3427 MWL), psi Credit for pressurizer spray, psi 11/-6 Effect of AMSAC on 19ne of turbine

+ 134/ + 165 trip and EFW flow ini. tion, psi Effect of iower EFW flow rate, psi

+ 59/ + 28 Peak pressures for Seabrook Station with 3074/3014 no mitigating equipment failures, psia Effect of loss of one EFW pump, psi

+ 18/ + 13 Peak pressures for Seabrook Station with 3092/3027 loss of one EFW pump, psia These peak pressures dernonstrate margin to the core damage limit of 3200 psig (or 3215 psia) discussed in Section 2.0, above.

5.0 MITIG ATING EOUIPMENT FAILURES IN ATWS RISK A probalistic evaluation of risk from ATWS events in SECY-83-293 provided the basis for the ATWS rule. An additional risk evaluation consistent with SECY-83-293 was performed by the Westinghouse Owners Group in WCAP-11993 (Reference 4) and provided to the NRC.

The contribution of failures in mitigating equipment to the overall ATWS risk is well documented in both these sources. WCAP 11993 concludes that Westinghouse PWRs exhibit ATWS core damage frequencies in compliance with the SECY-83 293 target. This is also true for Seabrook Station as documented in Reference 5 and submitted to the NRC by Reference 6.

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6.0 CONCLUSION

The above evaluation demonstrates that Seabrook Station is in compliance with the ATWS Rule,10CFR50.62 and the' Rule basis in SECY 83 293.

7.0 REFERENCES

1. Burnett. T.W.T., et.al., Westinchouse Antleinated Transients Without Trin Analysis.

WCAP 8330, Westinghouse Electric Corporation, August,1974 2.

Letter NS TMA 2182, Anderson, T.M. (Westinghouse Electric Corporation) to lianauer, S.it (USNRC), ATWS Submitta). December 30, 1979.

3. Dirks, W.J., Amendments to 16 CFR 50 Related to AntleinJted Transients Without Scram ( ATWS) Events. SECY 83 293, USNRC, July 19, 1983.

4. ' Sloane, B.D., et.al., Joint Westinchouse Owners Grouc / We@.chouse Procram:

Assessment of Comotinnee With ATWS Rule Basis for Westinnla use PWRs. WCAP-11993, Westinghouse Electric Corporation, December 1988.

5.

Kiper, K.L., et.al., Individ_ual Plant Examination Report for Seabrook Station.

Response to Generie Letter 88 20. NHY Engineering Report No. 90 01, March 1991.

6.

NHY Letter NYN 91034, Drawbridge, B.L. to USNRC, Sunnlementary Response to Generic Letter 88 20. March 1,1991.

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